Fuel element for nuclear reactors



April 24, 1962 Filed Sept. 17,- 1957 R. J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Shee'f l I I l l I l I I I l l I I I I I J INVENTOR REINHOLD J. BARGHET Arm/vo s April 1962 R. J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Sheet 2 Filed Sept. 17, 195'? INVENTOR REINHOLD J. BARCHET H L H 1.1., UW E3,

Arroe/v -ys April 24, 1962 R. J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Sheet 3 Filed Sept. 17, 1957 INVENTOR REINHOLD J. BARCHET A m/vars April 1962 R. J. BARCHET 3,031,388

FUEL ELEMENT FOR NUCLEAR REACTORS Filed Sept. 17, 1957 9 Sheets-Sheet 4 INVENTOR REINHOLD J.BARCHET Array/vans" April 24, 1962 R. J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Sheet 5 Filed Sept. 17, 1957 Snow Any/retranw W W w MUS-Was INVENTOR REINHOLD J. BARCHET rrows 5rd April 24, 1962 J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Sheet 6 Filed Sept. 17, 1957 INVENTOR REINHOLD J. BARCHET ,4 hug/vars April 24, 1962 R. J. BARCHET 3,031,388

7 FUEL ELEMENT FOR NUCLEAR REACTORS Filed Sept. 17, 1957 9 Sheets-Sheet 'r fc'urnra 72591144 FZUX 23456789vanilla/4151617161910 Ti j 14 (40/06 lzvcufs T a. l S

C /r/qu Hm gmoe ams- U 6 a lo I: '4 I6 INVENTOR @wuJ-fiVu/as REINHOLD J. BARCHET nrrapv rs April 24, 1962 R. J. BARCHET 3,031,388

FUEL ELEMENT FOR NUCLEAR REACTORS Filed Sept. 17, 1957 9 Sheets-Sheet 8 INVENTOR. fay/x010 t/ 54 966457- B (SQMYMCSI. (JQQLQ April 24, 1962 R. J. BARCHET FUEL ELEMENT FOR NUCLEAR REACTORS 9 Sheets-Sheet 9 Filed Sept. 17, 1957 3,031,388 FUEL ELEMENT FOR NUCLEAR REACTORS Reinhold J. Barchet, Hagerstown, Md., assignor to Martin Marietta Corporation, a corporation of Maryland Filed Sept. 17, 1957, Ser. No. 685,058

10 Claims. (Ci. 204-1541) 7 The present invention relates generally to nuclear power plants, and more particularly to a heterogeneous, pressurized, light-water moderated and cooled reactor system which may be readily transported and installed and which contains tubular fuel elements arranged in such a manner as to effect even flow distribution of a coolant between the internal and external surfaces thereof.

Small reactor power packages ranging up to 20,000 kilowatts of electrical output will become increasingly important in supplying power to many small and yetto-be industrialized areas of the world. The need also exists for packaged power components which may be carried by air to a military site and there connected together to form a complete system.

Conventional sources of heat, light and power present diificult logistic problems for military and industrial operations in many parts of the world where ordinary fuel supplies are not available. A nuclear power plant, on the other hand, requires no fuel storage, it practically eliminates fuel supply problems and demands a of operating personnel and maintenance.

In view of the foregoing it is the principal object of the present invention to provide an efiicient nuclearpower plant of relatively small size and light weight which may be readily transported from the point of manufac ture to the site of operation.

It is also an object of the invention to provide a nuclear power plant of inherently simple design which is nevertheless extremely safe and reliable in operation. A power system in accordance with the present invention may be easily installed and prepared for operation.

, The key component in a power generating system in accordance with the invention is the nuclear-power reactor which may, if desired, be optimized for light weight rather than thermal efliciency. The reactor is of the pressurized Water type but no thermal shielding is provided within the pressure vessel, thus permitting a substantial reduction in the overall diameter of the reactor. This in turn allows a reduction in wall thickness and hence a reduction in the amount of heat generated as compared to a heavier walled vessel. The thermal shielding and nuclear reflection are accomplished by the pressure shell, its outer layer of insulation and the surrounding water in which the entire reactor is submerged.

' The second principal factor that has made possible a reduction in size as well as a considerable simplification of the core itself is the basic fuel element employed.

Ideally, the fuel element should posses good radiation stability and be able to withstand high temperatures. Its construction should be such as to provide the largest possible heat transfer area to facilitate the transmittal of heat generated therein. In addition, it should be relatively easy and inexpensive to fabricate. In the design of United States Patent 3,031,388 Patented Apr. 24, 1962 ice weight of the core but also act to increase the metal to water ratio, thereby increasing the neutron-absorbing cross-section and reducing the eificiency of the device.

In accordance with the present invention the core of the reactor is constituted of fuel elements of tubular geometry having improved structural strength and rigidity. No supporting structure is needed within the nuclear core thereby greatly enhancing the nuclear efficiency. The tubular construction lends itself to the how of Water both inside and outside of the tubes to afi'ord a maximum of heat transfer surface for a given volume of heat generating material. This basic shape also gives excellent hydrodynamic characteristics. All these factors combine to provide a compact core of size for the required performance.

It is a further object of the invention to provide a nuclear core constituted by an assembly of fuel element bundles fitted into the core configuration, each bundle having three extended tubes held together by an adaptor ring which fits into a socket built into a supporting grid. Through the usage of tubular fuel elements it is possible to eliminate all obstructing supports within the efiective heat transfer surface area as individual envelopes for each fuel unit are omitted, resulting in the lowest possible metal to water ratio.

It is a primary object of the invention to provide an improved structural design for a tubular fuel element and an arrangement for interconnecting a group of such elements into a bundle whereby the equivalent cross-section 1 on the outside of the tubes is substantially equal to that fuel elements for portable pack-aged power stations, lightbeing determined of the inside cross-section thereof to cause a properdistribution of coolant flow therein.

More vparticularly it is an object of the invention to provide a tubular fuel element whose ends are flared and fluted to effect the desired flow distribution.

Also an object of the invention is to provide a bundle of interconnected tubular elements as above described which are held together in a geometrical array by spot welding. The bundles may be symmetrically clustered into a strong and simple assembly withno added fixtures or other structural connectors, the weight of the bundles solely by the weight of the constituent tubes.

For a better understanding of the invention, as well as further objects and features thereof, reference is made to the following detailed description of one specific embodiment thereof which illustrates the invention; The description is to be read in connection with the accompanying drawings wherein like elements in the views are identified by like reference numerals.

In the drawings:

FIGURE 1 is a schematic view showing the primary loop fiow diagram for an embodiment of the present invention;

FIG. 2 is a perspective view of a primary loop for a reactor system in accordance with the invention, the casing being cut away to expose the interior;

FIG. .3 is a perspective view of the nuclear reactor forming part of the primary loop, the pressure vessel being cut away to reveal the core and the control rods;' FIG. 4 is a perspective view of a single tubular fuel element;

FIG. 5 is a schematic showing of the construction of the live portion of the fuel element;

FIG. 6 illustrates one embodiment of a bundle constituted by a plurality of interconnected tubular fuel elements;

FIG. 7 is a perspective View of the reactor core constituted by an assemblage of fuel element bundles;

FIG. 8 is a schematic diagram of the reactor core;

FIG. 9 is a graph showing the relative worth of control rods versus position;

FIG. 10 is a longitudinal sectional view of the actuating mechanism for a control rod;

FIGS. 11 and 12 are schematic diagrams'of the power reactor;

FIG. 13 is a graph showing the relative fast flux versus radius for the reactor;

FIG. 14 is a graph showing the relative thermal flux versus radius of the reactor;

FIG. 15 is a graph showing the critical mass versus the radius of the reactor;

FIG. 16 is a perspective view of a tubular fuel element before the ends are belled;

FIG. 17 is a perspective view of one of the end portions of the tubes after it is formed in accordance with the in vention;

FIG. 18 is a longitudinal section taken through the tube end portion;

FIG. 19 is an elevation view of a group of interconnected tubes in accordance with the invention;

FIG. 20 is an end view of the tubes shown in FIG. 19;

FIG. 21 is a perspective view of an embodiment of a bundle of tube elements; and

FIG. 22 is an end view in schematic form showing an alternative form of a bundle of tubular fuel elements.

GENERAL DESCRIPTION Referring now to FIGS. 1 and 2 there is shown the primary loop of a power reactor system in accordance with one embodiment of the invention. The primary loop represents one component in a system which also includes a turbo-generator driven by steam produced in the primary loop, suitable switch gear deaerators, etc. The invention resides in the structure and function of the primary loop, hence the turbo-generator and other conventional components which receive this power from the primary loop will not be shown herein. The principle of the reactor system is the same as that of a conventional steam-turbine generating plant, the source of heat being the principal difference.

The major elements of the primary loop are a steel envelope 10 within which is contained a nuclear reactor generally designated by numeral 11. The steel envelope is preferably surrounded by reinforced concrete shielding 12. A heat exchanger or steam generator 13 is provided in association with the reactor 11, the output pipe 14 of the generator being the steam line to the turbine or other steam driven motor for producing power. Canned pumps 15 furnish pressurized water to the reactor, the primary piping being designated by numeral 16. The interior of the steel envelope 10 is filled with shielding water 17. The power reactor is designed without a complete thermal and radiation shield and when installed it is placed below ground and preferably in a larger body of water to take advantage of natural shielding.

By elimination of the thermal shield and use of a small pressure vessel, the reactor weight has been reduced by as high a factor as percent.

Referring now to FIG. 3 there is separately shown the nuclear reactor 11, the reactor comprising a core 18 of tubular fuel elements, supported at either end by upper and lower grids 19 and 19a within a pressure vessel 20. Vessel 20 is encased in an insulation jacket 21. Pressurized Water is fed into the pressure vessel 20 through inlet pipe 22 and diffused therein before passage through the core 18 by means of baffle plates 23. Bafile plates 23 are perforated to create an even velocity profile.

The heated water is discharged from the reactor through the primary line 16. Control of the reactor is effected by 4- control rods 24 which are projected into the core, or retracted therefrom by means of pistons 24a operated by a control rod mechanism 25. Four control rods are provided, one in the center and three eccentrically thereof, the rods being fabricated of Boron-10.

The reactor shown in FIG. 3 is designed to operate at a power output level (heat) of 8 megawatts. The performance and design data of such a device may, for example, be as follows:

Reactor Weight (minus shield and water) 6000 pounds. Overall dimensions at pressure vessel 84 inches long by 24% inches outside diameter and 23 inches inside diameter. Core life 550 days. Reactor coolant 2000 p.s.i. water. Reactor coolant flow 1,295,000 pounds per hour. Coolant velocity in core 3.6 ft./sec. Mean coolant temperature 510 Fahrenheit. Coolant temperature drop 18 Fahrenheit. Metal/water ratio 0.195.

The nuclear data may, for example, be as follows:

Critical mass 17.66 kilograms of U Burn-up (550 full power days) 31 percent.

Average flux (thermal) 1.3 10 neutrons per (cm?) (see). Burnable poison loading 125.9 grams of natural boron. Control rods:

Number Four. Material 3.5% boron-10 in stainless steel.

Length (active) 23 inches.

Diameter 3 inches.

Travel (active) 23 inches.

An individual fuel element 26 is shown in FIG. 4. Each fuel element 26 may be constituted by an 0.375 inch outside diameter tube with a 0.030 inch wall. As shown in FIGS. 5 and 16 the matrix 27 at the fuel element is formed by a sintered mixture of stainless steel powder, uranium oxide and boron nitride (RN) and is 0.020 inch thick. The composition of this matrix by weight may be 78.75% stainless steel, 21% fully enriched U0 and 0.25% BN. Preferably this matrix is clad with a 0.005 inch stainless steel outer cladding 28 and a similar inner cladding 29 on the inner surface, thereby making the total wall thickness 0.030 inch.

The unfueled ends 26a and 26b of each fuel element are flared to an outside diameter equal to the desired center-to-center spacing of the tubes. Flute type switches 26c and 26d are formed in the flared ends to effect equal flow distribution of the coolant between the inner and outer surfaces of the tube.

As shown in FIGS. 6 and 21 a group of tubular fuel elements 26 may be spot welded together at their belled ends to form a bundle 30 generally hexagonal in shape. As shown in FIG. 6 three tubes 26x may be extended at the center of the bundle so that they project from either end of the bundle to provide means for holding the bundles in position in the core and means for grasping the bundles to remove them as desired. The active fuel area of the extended length tubes 26x is identical to that of the shorter tubes, the dead end extensions being solely to facilitate handling and securing the bundles. A ring 31 may be placed over the ends of the extended tubes at either end of the bundle.

To form the core assembly, a predetermined number of bundles are grouped together geometrically as shown in PEG. 7, with four spaces left vacant to make room for for axial movement by the pistons 24a.

3) is provided at the points of intersection with sockets 33 which accommodate the end rings 31 on the bundles, thev grid serving to align and retain the core in a fixed position within the reactor. .The control rods 24 are reciprocable Within the well sleeves 32 and are supported Below and forming apart of each control rod assembly 24 (see FIG.

3) there may be a small bundle 34 of twenty-seven fuel tubes. As the. control rod assembly is withdrawn by the piston 24a the small fuel bundle 34 is drawn into the core. This is done to reduce neutron flux peaking in the control rod wells when rods are withdrawn. The rod scram mechanism, to be later described, is designed to scram in 0.3 second. The control problem is minimized by the inherently large negative temperature coefiicient of the reactor and the use of burnable poisons in the fuel matrix. As stated the use of both neutron obsorbing and fuel materials keeps the flux from peaking when the control rods, are not in the reactor and increases the effectiveness of the control rods by reducing fuel content when the control rods are in place. The value of the present type of control rods in the reactor without the added factor of the fuel on the ends is 35.5% for the entire cluster in the cold clean case, and 21% for the entire cluster in the hot, clean case. These values were computed under conditions designed to yield the most conservative results. The reactor has twice as much control as is needed at all times. Any one rod will scram the reactor, and if all of the others should fail to scram, the reversal of' just one of these rods will hold the reactor sub-critical. For the value of the rods versus inches of withdrawal or insertion see FIG. 9. e

As pointed out previously there are 27 fuel tubes attached to the base of each control rod. The amount of fuel which is removed by complete insertion of all rods is 6.98%.

Control Rod Actuation The control rodactuating mechanism (25 in FIG. 3) must be .capable of moving either up or down at a controlled rate of speed. The movement and location of the rods is. vitally important. Inherent safety features are necessary in order to scram the reactor. In theevent of power failure or for nuclear reasons the rods must be driven to a safe position in an extremely short time.

The control rod actuator is shown inFIG. l0 and comprises a motor 35, a solenoid 36, a ball bearing screw jack 37 and scram springs 38. p

The motor 35 is essentially of the vertical, flangemounted, canned, constant speed, polyphase squirrelcage induction type. The motor is attached to the pressure vessel by means of a bolted flange fitted with O ring seals. The input is 3-phase, 440 volts, 60-cycle alternating current. The synchronous speed is 3600 revolutions per minute, with an output of hi horsepower. It is also reversible.

The solenoid which is used on the actuator is a vertical screw-mounted, push-down type of canned solenoid. It

is attached to the pressure vessel by means of a threaded shoulder fitted with a wedge-type sealing lip. Ratings of the solenoid are: input, 2.8 volts nominal direct current battery source, stroke 0.375 inch, nominal spring load ten pounds of force initially, increasing to 30- pounds of force dead weight in the seat position.

The canned feature of the motor and solenoid is merely a process which leaves the rotor of the motor and the plunger of the solenoid in the high temperature, high pressure zone, while the windings are sealed away from this condition.

-turn; therefore, the screw climbs up the square shaft. 7

The ball-bearing screw jack 37 is a device, which as the name implies, is a nut 37a and screw 3711, which can be driven in either direction through a series of balls. The balls after completing a driving cycle continue through a circuit and return for another driving phase.

Two sets of gears 39 and 40 provide the necessary reduction from the motor to the ball-bearing screw jack. The first set are called spiroid gears and give an extremely high reduction. The second set are typical bevel gears and have a much smaller reduction.

Three proximityatype limit switches (not shown) are used on the unit. These switches are magnets which are located outside the pressure vessel. For the most part the parts are of non-magnetic stainless steel. However, in order to actuate the limit switches a magnetic material is used. One switch indicates when the rod is in its lowest or safe position. Another switch tells when the rod is completely withdrawn. The third switch is used to form a part of the latching mechanism.

To clarify the operation of the mechanism, it can be assumed that the control rod is in the scrammed or safe position. The motor 35 is then started. The square shaft 41 which is driven through the gear train by the motor now rotates. The square shaft in turn is now a driving member rotating the screw 37b. The nut 37a for this screw is at the lowest position and keyed so it cannot When the screw reaches the upper limit of its travel one of the limit switches stops the motor and energizes the solenoid. The solenoid 36 in turn closes three trip levers 36a which grasp the top of the screw.

The motor direction is-now reversed and the square rod causes the screw to rotate again. In this instance the nut, still keyed in place, climbs up the screw until the desired position is reached. As the nut rises it compresses the scram springs 33. The unit is now cooked and ready to perform the scram function if required.

In the event that a scram becomes necessary or a power failure occurs the solenoid becomes deenergized. The latching levers 36a are now permitted to open and the scram springs drive the rod plus the nut and screw home to a safe position.

When the complete unit reaches the lower end of its travel it enters a dash pot. The dash pot is merely a cushion, mounted near the top of thelpressure vessel, to absorb the shock of decelerating this rapidly moving mass.

Core Design and Flow Analysis Flow through bundless.-As indicated previously, the coolant flows inside and outside the fuel element tubes. To obtain optimum heat transfer the bulk flow velocity and bulk temperature rise has to be the same inside and outside the tubes. These conditions are induced if the equivalent cross-section of the flow path outside of the tubes is equal to the inside cross-section of the tubes.

The reason that this spacing is the optimum can be shown as follows:

Consider the case in which pressurized Water is flowing both inside and outside a bundle of tubes. The tubes are spaced such that the equivalent cross-section of the outside flow passages is the same as the inside cross-section of the tubes.

We know that the pressure drops for the flow inside and outside the tube are equal because the paths are parallel. We can also assume that the inside and outside tube wall temperatures at any given points are equal, because the resistance of the tube wall is small compared to that of the water film. It is desirable that the coolant temperature at corresponding points inside and outside be the same, since otherwise there would be a mixing and averaging of the different temperature water when the interior and exterior flows met at the exit of the tube. Any boiling is contrary to the objective of the pressurized water reactor.

For turbulent flow the pressure drop inside and outside the tube is given by the Fanning equation:

a e ULKi Marga 29 a "-1121 (2) D1-2 2g 2 Since Then because the physical properties of the water and the equivalent diameter are the same on both sides of the tube.

then

since The ratio of heat transferred from the inside surface to that, transferred from the outside surface is given by the following two equations:

(2) Egfiwc anfmiv c mfig But the free flow area, from the definition of equivalent diameter, is given by the equation so that which is necessary from the Equation 1 directly above.

The above argument shows, then, that only if the tube spacing is such that the equivalent cross-section of the outside flow area is equal to the inside cross-section of the tube will the coolant temperature at corresponding points inside and outside the tube be equal.

Referring now to FIGS. 5 and 16, there is shown a tubular fuel element before the end portions thereof are flared. The element comprises a tubular core 27 including fissionable particles contained in a matrix, the core being surrounded by an outer cladding tube 28 whose interior face is bonded to the exterior surface of the core. Disposed concentrically within the core is an inner cladding tube 29 whose exterior face is bounded to the interior surface of the core. In the end portions 26a and 26b of the element, the core contains no active fuel. The end portions therefore are dead to seal the element.

As shown in FIGS. 4, 17, 18 and 19, the end portions are expanded to a diameter equal to the desired centerto-center spacing of the tubes. Thus end portions 26a and 26b are constituted by an enlarged collar section 26 of say a half inch length and tapered section 26g having for example a 12 degree incline, which tapered section iirerges with the original main portion of the tubular elements. The expansion may be effected by conventional expansion or swaging processes. Tolerance build-up is eliminated through the use of a sizing die to ensure DUO-.002 inch tolerance.

To obtain evenly divided flow between the inside and outside of the tube, flutes 26c and 26d which may be diametrically opposed as shown, are depressed into the inner stream. These flutes may be fabricated with a punch which shears and depresses the wall in one operation, the depths of the flutes being determined by flow require ments.

In a specific embodiment of the invention the desired evenly divided flow between the inside and the outside of the tube, having an outer diameter of 0.375" and an inner diameter of 0.315", has been obtained by having a onehalf inch enlarged or flared collar section 26 at the end portions 26a and 26b of the tube with tapered sections 26g intermediate the collar sections 26 and the body of the tube having a 12 degree incline and a one-quarter inch length. In this embodiment the centers of these flutes at each end of the tube may be annularly displaced degrees from each other, thus producing three flutes instead of the two flutes as shown, with each of these flutes having a depth of 0.150 inch.

A plurality of these tubes having flared end portions 26a and 26b having an inner diameter of 0.406 inch and an outer diameter of 0.466 inch may be spot welded together at the tangential points on these flared end collar sections at which the adjoining tubes touch to form a bundle 30 as shown in FIGS. 6, 20, 21 and 22. This bundle is an inherently rigid and strong structure which is self-supporting and in which the spacing of the tubes is determined by the diameter of the enlarged end portions, thereby dispensing with spacing grids or other expedients adapted to separate and hold the tubes in position. Further, with the spot welds securing adjoining tubes together positioned at the end portions 26a and 26b of the tubes, the possibility of hot spots is avoided. The reason for this is that the spot welds join the dead end materials and there is no chance of burn through of fissionable ma erial as would be the case if the spot welds were positioned along the live or active part of the tube. Still further, because of this positioning of the spot welds the fabricator and welder of the bundles does not have to be as careful in his welding procedures because there is no chance of burn-through.

As shown in FIGS. 6 and 21 the tubes may be joined together to form a bundle having a hexagonal configuration. A group of such bundles may be clustered together to form the core of the reactor.

In the tubes illustrated in FIGS. 17-21, a pair of diametrically opposed flutes are formed in the end portions. Alternatively, as shown in FIG. 22, three symmetrically arranged flutes 26c and 26d and 26a may be cut into the 9 end portions to obtain an even flow distribution between the inner and outer surfaces. 'This'alternative arrangement of three symmetrically arranged flutes 26c, 26d and 26e is the arrangement referred to in the specific example above.

Flow outside core-The core, as seen in FIGS. 11 and 12, is in the approximate form of a right circular cylinder. The perimeter of this cylinder is irregular, its greatest diameter being 21.5 inches and its mean diameter 20.75 inches. This core fits into the cylindrical pressure vessel which has aninside diameter of 23.0 inches. An annular area between the pressure vessel and the core of about -3 square inches exists through which the coolant flow must be restrained.

An annular can was chosen to restrain this flow. The

can is as long as the core itself and has its inside perimeter so shaped as to fit the irregular perimeter of the core;

To permit cooling of the pressure vessel wall, the'outside' diameter of the can was set at 22.875 inches, which is 0.125 inch less than the inside diameter of the pressure vessel. A V inch wide annulus then exists for flow between the can and the pressure vessel. Calculations reveal that 24 gallons per minute of coolant will flow through this annular fiow area at a velocity of two feet per second. W

The total heat generated in the pressure vessel wall has been estimated to be 282,000 B.t.u. per hour or 82.6 kilowatts. If it is assumed that the 'outside of the pressure vessel is perfectly insulated, themaximum heat flux will occur at the inner surface of the wall and will be about 5000 B.t.u. per hour foot squared.

With these heat generation data and the flow data, the following temperatures were calculated:

Degrees Fahrenheit Annul-us inlet temperature 501 Core inlet temperature 501 Coolant temperature rise through annulus 9 Coolant temperature at midpoint 50 5.5 Film temperature rise at midpoint 32 Maximum wall temperature at midpoint 537.5

heat raised the coolant temperature 1.5 degrees Fahrenheit. 7

Flow around control rods-Four control rods, as

shown in FIG. ll, pierce the core. One is located on the center line and the remaining three are equally spaced on a 13.4-inch diameter circle. upper poison section in the form of a hexagonal tube 26 Each rod consists of an and A; inches long ,(23 inch active) and a lower section which is a hexagonalcluster of 27 fuel tubes. 25 and /2 inches long. 'The entire rod is surrounded by a hexagonal sheath or can 53 and /8 inches long. A clearance of inch exists between the control rod and the can.

Three flow paths were investigated for the control rod assembly. The first path is through the tube' bundle and then through the inside of the poison section. The

second is in the annular area between the fuel bundle, poison section and the enclosing can. The third path is the annular area between the outside of the can and the surrounding tube bundles.

The resistance to flow for the two paths inside the enclosing can depends on the location of the control rods.

.Each of these paths was investigated for three positions ofthe control rods-fully inserted, withdrawn half-way and fully withdrawn. 7,

Equivalent diameters for the various" paths were calculated and flows found from a trial and error balancing .of resistance and known pressure drops.

Table 1 summarizes the results for one rod.

TABLE 1.FLOW DISTRIBUTIONS Fully Halfway Fully Inserted Withdrawn Withdrawn U -l Q, V2 Q Q Sec. g.m.p.

Path 1:

Inside Tube bundle 3. 7 3. 7 3. 7 Inside poison 36. 2 36. 2 36. 2

e ion 1. 9 1. 9 1. 9 Path 2:

Outside Tube bundle l. 6 2.1 3. 4 Outside poison 6.95 9. l 14. 7

eotion 2. l 2. 7 0 Path 3: Outside Sheath 3. 0 3. 0 3. 0

1 Velocity decrease is balanced by areaincrease, so flow per tube outside sheath remains the same as for the regular full tube or about 2.13 gallons per minute per tube. r r

2 Velocity in feet per second.

5 Flow in gallons per minute.

Total fl0w.--The total flow through the core with the control rods in the probable operating positions, i.e., the three outer rods fully withdrawn and the central rod half Withdrawn, is estimated to be 3316 gallons per minute.

Of this total, 3065 gallons per minute will be flowing through and around the fuel tubes and the remainder will flow outside the core and through the control rods.

Flow at inlet-To effect a smooth transition from the 8-inch diameter inlet pipe to the 22 inch diameter reactor .vessel, a difiuser core with four perforated plates or screens is to be used (see plates 23 in FIG. 3).

The four plates are of uniform solidity ratio -0.4-

.and are so spaced in the diifuser core that the pressure .drop across each one is equal to the rise in static pressure, or loss in velocity pressure in the space preceding it. The total pressure drop was calculated to be about three pounds per square inch.

Pressure dr0ps.-The pressure drops through the various components of the primary loop were calculated to be as follows:

Component; Pressure drop (p.si.) Inlet flow control screens .9 Reactor vessel 1.6

Steam generator 4.4 Piping 3.8

Total 12.7

Reactor Vessel The design of the reactor vessel must not only take into account the stresses due to internal pressure, structural support, earthquakes, etc., but also the extremely high thermal stresses which are encountered due to heat generation in the pressure vessel walls. It has been found from a comparison of material manufacturing and fabrisection with 28 high-strength one and one-quarter inch diameter bolts. Four bosses, with two inch holes for the control rod shafts, are welded to the head. An eight- .inch diameter inletand outlet are the only openings in the vessel body. Brackets are welded to the inside of the vessel walls for holding the reactor core in place.

1 l The following tabulation is a description of the pressure vessel:

Pressure vessel:

(maximum shear). Thickness of heads 1.5 inches. Semi-ellipsoidal heads ASAstandard.

Flange thickness 3 inches. Bolt circle diameter 30 inches. Size of bolts 1.25 inches. Number of bolts 28. Type of bolts 12 Unified National Fine (160,000 p.s.i. minimum). Overall length of vessel 90% inches. Inlet opening 8 inches. Outlet opening 8 inches. Insulation Aerogel.

Fission product poisoning-One of the main requirements for automatic control is to overcome the xenon transient. Enough excess reactivity must be provided so that the reactor can be started at any time. It is also necessary to determine the rate of change of the xenon concentration so that the control rods can be designed accordingly.

While the reactor of the described embodiment is in operation, the xenon concentration increases toward an equilibrium value which occurs approximately 72 hours after start-up, although a very close value is reached in about twenty-four hours. For the reactor described herein, the equilibrium valve attained by the xenon concentration is 1.73l 10 atoms per cubic centimeter. At this point the rate of formation of xenon is equal to the rate of removal. If the reactor is in continuous operation, the amount of reactivity loss to be overcome is the amount calculated at the equilibrium point (2.5 percent AK per K).

It for some reason the reactor should shutdown, the xenon concentration increases steadily for several hours before reaching a maximum because it continues to be formed as a decay product. After reaching the maximum it slowly decreases. This xenon hump occurs about eight hours after shutdown. The xenon concentration at maximum build-up point is 2.6939x10 atoms per cubic cen' timeter, almost twice the amount of that at equilibrium. However, this is a low value in comparison to those of reactors with fluxes in the order of 10 and would not seriously hinder the reactor should it be started up at this point. The excess K needed to overcome the loss of reactivity is only 3.9 percent, whereas for a flux of 10 as much as 35 percent reactivity can be tied up by xenon build-up.

If the reactor is started up at or near the maximum build-up point, the xenon concentration will decrease very rapidly, falling below the equilibrium value. It will then gradually work its way up again. It is here that the rate of change of the xenon concentration is greatest causing a gain in reactivity at the rate 2.579 10- AK per second immediately after start-up. The amount of reactivity per second which can be taken care of by the control rods is 133 X10 which is more than adequate to control the reactivity drop due to xenon burn-out after start-up. The results of calculations show that the reactor fuel loading and multiplication factor keep the xenon transient under control at all times.

From these calculations, it can be seen that xenon poses no problem in either the operation of the reactor or in automatic control of the system. At its highest point, the xenon concentration is not large enough to make the reactor sub-critical. This indicates that if the reactor should be shut down, no Waiting period is necessary to 12 start it again. These results also shown that should the reactor be started again at the highest xenon concentration, the danger of having the rate of poisoning go down faster than control rods could absorb the resulting excess reactivity would not exist. Under present conditions fission produce poisoning does not affect the stability of the system.

Long Term Reactivity Studies Long term reactivity study is the name given to a group of problems whose solution involves the entire lifetime of the reactor. Under the heading of long term reactivity are such things as determination of the loading of the core, control rod size, negative temperature coefficient, burnable poisons in the reactor and fission product poisoning. Having an estimate of the excess reactivity required in the reactor the control rod size can be determined. The amount of fuel necessary to keep the reactor going for its entire lifetime determines the fuel loading which in turn dictates the burnable poison content in the system. This is a summary of the completed work on long term reactivity studies for the reactor described herein.

Total heat production 8 megawatts.

Core lifetime 1 /2 years.

Core temperature-cold 68 degrees Fahrenheit. Core temperature-hot 510 degrees Fahrenheit.

Assumptions made for the present work are:

(1) The core is a homogeneous mixture of the components. This should be a fairly accurate assumption for this particular core design and study.

(2) Fuel is fully enriched U (3) Fast fission factor and the resonance escape probability product equals one-e-p=l;

(4) Macroscopic transport cross-section of U is small enough to be negligible;

(5) Burn-up is based on 1.25 gram per megawatt day;

(6) Transport cross-secti0ns and Fermi Age do not change due to build-up of fission products;

(7) Samarium poisoning reaches equilibrium at the same time as xenon poisoning;

(8) Neutron temperatures are equal to the mean coolant temperature.

It is advantageous to have a flat, long term reactivity curve. This is best accomplished by having the burnable poison burn-up similar to the fuel burn-up. The crosssection of the burnable poison should be close to the cross-section of the fuel.

On the basis of the physical characteristics of various poison materials and the burnable poison requirements boron-10 was chosen as most suitable for this reactor.

The abundance of boron-l0 in natural boron is 18.8 percent. Since boron-l0 has such a high cross-section relative to the other isotopes, only the boron-l0 need be considered as a neutron absorber if the reactor core is to be poisoned with natural boron.

Since the excess reactivity decreases quite rapidly until xenon has reached its equilibrium concentration after start-up (the most rapid change is due to the temperature effect) and the increase in excess reactivity due to the depletion of boron-l0 is in comparison slow, the obvious reference point for the consideration of burnable poison is at about thirty hours when both the temperature and steady state fission products are in equilibrium. Choosing this reference point and arbitrarily choosing 0.0075 as the excess reactivity to be available at this point after 13 introducing the boron, the amount of boron needed is found by adjusting the thermal utilization and leakage until Since the boron burns out exponentially, and the fuel burns up more or less linearly, the maximum reactivity in the system occurs after 250 days of operation.

Based on 1.25 grams per megawatt day burn-up, the amount of U to be burned up to give 12 megawatt years of operation is 5.475 kilograms. The critical mass calculated for the hot clean reactor is 8.56 kilograms. The I With the imposed "31 percent U burn-up limit, the

system contains an excess of 3'.66 kilograms at shutdown (the core loading at 3 1 percent-U burn-up being 17.6612 kilograms). The excess fuel at shutdown-is 0.01 (maximum) in terms of excess reactivity.

To find the maximum amount of external control needed, it was assumed that the reactor has been operating for about 250 days and was shut down long enough for the xenon to decay out of the system. The maximum negative reactivity needed to hold the reactor sub-critical was about'tllll.

Consideration was given also to the case Where the reactor is shut down and a start-up is required Within a few hours after shutdownJ The usual problem in such a case is an override of the xenon which reaches a maximum concentration six to nine hours after shutdown, at which time the xenon decay overtakes the xenon production (iodine decay). In the reactor described herein, with a flux of about 156x10 neutrons per centimeter squared per second at v31 percent U bur11-up, the increase in xenon poisoning after shutdown is so small that it. may be considered to stay at the equilibrium concentration attained during operation. It remains at this equilibrium until the decay-becomes the dominant effect at about nine hours after shutdown.

While there has been shown what is considered to be a preferred embodiment of a reactor in accordance with the invention, it will be recognized that many changes and modifications may be made therein without departing from the essential features of the invention. It is intended, therefore, in the appended claims to cover all such changes as fall within the true spirit of the invention. For convenience and requirements of adequate disclosure, there has been disclosed herein certain inventions which are or will be the subject of copending applications filed by other applicants who are employees of applicants employer, The Martin Company.

What is claimed is: t

1. An integral fuel element for a nuclear reactor comprising a clad tubular section containing fissionable material, a collar adjoining each end of said tubular section having a diameter greater than the diameter of the tubular section, and a flared section intermediateeach end of the tubular section and each said collar, said flared sections having a pluraliy of flutes depressed therein, in circumferentially equispaced relation whereby coolant fluid may be directed by said flutes over the exterior of said section.

14 ,2. The nuclear reactor fuel element of claim 1 wherein the flutes are depressed for causing an axial flow of fluids through the collar and flared section to be evenly divided between the inside and the outside of the tubular section. 3. The nuclear reactor fuel element of claim 1 wherein said collars and said flared sections are formed from stainless steel.

4. The nuclear reactor fuel element of claim 1 wherein said collars and said flared sections are formed from Stainless steel and said tubular section is clad on its inner and outer surfaces with concentric stainless steel tubes. 5. A nuclear reactor fuel element comprising a clad ubular section having an inner diameter of 0.315 inch,

'an outer diameter of 0.375 inch and containing fissionable material, a collar adjoining each end of said tubular element having a length of one-half inch, an inner diameter of 0.405 inch and an outer diameter of 0.466 inch,

a flared section intermediate each end of the tubular e lement and each said collar having a length of one-quarter inch and three flutes depressed in each flared section annularly displaced from each other, each of said flutes having a depth of 0.150 inch.

r 6; The'fuel element of claim 1 wherein said flutes define, a flow path outside said tubular section having an equivalent cross-section equal to the internal cross sectiomof said tubular section, whereby the temperature of coolant fluid flowing at corresponding points inside and outside said tubular section is the same. 0

7; A tubular element for a nuclear reactor comprising a tubular core of fissionable material encased in inner and outer metal tubes, said metal tubes extending beyond the ends of said tubular core so as to provide deadends, said dead ends being of larger diameter than the portion of said elementintermediate thereof and containing flow passages therethrough for the flow of coolant through said inner tube as Well as around said outer tube.

8. A fuel element arrangement for a nuclear reactor comprising a plurality of substantially tubular elements, each said element comprising a tubular core of fissionable material encased in inner and outer metal tubes, said metal tubes extending beyond the ends of said tubular core so as to provide dead ends, said dead ends being of larger diameter than the portion of said element intermediate thereof and containing flow passages therethrough for the flow of coolant through said inner tube and around said outer tube, said elements being disposed in substantially parallel relation in a bundle, said passages being of a size to divide substantially evenly the flow of fluid inside and outside said fuel elements.

9. The method of providing cooling for a bundle of tubular elements in a nuclear reactor core comprising the steps of assembling enlarged fluted ends upon each fuel element, placing said elements in a substantially parallel array, and causing coolant fluid to circulate through and around said elements, said fluted ends proportioning the flow of coolant fluid around the exterior of said elements.

10. The method of manufacturing an improved tubular fuel element comprising the steps of securing enlarged ends upon a tubular section containing fissionable material, and fluting each of said ends to provide a fluid path for coolant about the exterior of said section. 

1. AN INTEGRAL FUEL ELEMENT FOR A NUCLEAR REACTOR COMPRISING A CLAD TUBULAR SECTION CONTAINING FISSIONABLE MATERIAL, A COLLOR ADJOINING EACH END OF SAID TUBULAR SECTION HAVING A DIAMETER GREATER THAN THE DIAMETER OF THE TUBULAR SECTION, AND A FLARED SECTION INTERMEDIATE EACH END OF THE TUBULAR SECTION AND EACH SAID COLLAR, SAID FLARED SECTION HAVING A PLURALITY OF FLUTES DEPRESSED THEREIN, IN CIRCUMFERENTIALLY EQUISPACED RELATION WHEREBY COOLANT FLUID MAY BE DIRECTED BY SAID FLUTES OVER THE EXTERIOR OF SAID SECTION.
 7. A TUBULAR ELEMENT FOR A NUCLEAR REACTOR COMPRISING A TUBULAR CORE OF FISSIONABLE MATERIAL ENCASED IN INNER AND OUTER METAL TUBES, SAID METAL TUBES EXTENDING BEYOND THE ENDS OF SAID TUBULAR CORE SO AS TO PROVIDE DEAD ENDS, SAID DEAD ENDS BEING OF LARGER DIAMETER THAN THE PORTION OF SAID ELEMENT INTERMEDIATE THEREOF AND CONTAINING FLOW PASSAGES THERETHROUGH FOR THE FLOW OF COOLANT THROUGH SAID INNER TUBE AS WELL AS AROUND SAID OUTER TUBES.
 9. THE METHOD OF PROVIDING COOLING FOR A BUNDLE OF TUBULAR ELEMENTS IN A NUCLEAR REACTOR CORE COMPRISING THE STEPS ASSEMBLING ENLARGE FLUTED ENDS UPON EACH FUEL ELEMENT, PLACING SAID ELEMENTS IN A SUBSTANTIALLY PARALLEL ARRAY, AND CAUSING COOLANT FLUID TO CIRCULATE THROUGH AND AROUND SAID ELEMENTS, SAID FLUTED ENDS PROPORTIONING THE FLOW OF COOLANT FLUID AROUND THE EXTERIOR OF SAID ELEMENTS.
 10. THE METHOD OF MANUFACTURING AN IMPROVED TUBULAR FUEL ELEMENT COMPRISING THE STEPS OF SECURING ENLARGED ENDS UPON A TUBULAR SECTION CONTAINING FISSIONABLE MATERIAL, AND FLUTING EACH OF SAID ENDS TO PROVIDE A FLUID PATH FOR COOLANT ABOUT THE EXTERIOR OF SAID SECTION. 